JPH0364427A - Corrosion resistant zirconium alloy containing copper, nickel and iron - Google Patents

Abstract

PURPOSE: To prevent reaction of nuclear fuel rods with cooling water and moderators and the occurrence of trouble in a nuclear reactor by leakage of radioactive fission products by using a Zr alloy having a specific compsn. as coating materials of the nuclear fuel of the reactor, thereby improving uniform corrosion resistance and nodal corrosion resistance.

CONSTITUTION: Many pieces of the fuel elements 14 formed by forming the nuclear fuel rods 16 consisting of UO₂, PuO₂, etc., at centers and coating the outer side thereof with tubes 21 made of the Zr alloy contg., by weight%, 0.5 to 2.0% Sn, 0.24 to 0.40% (more particularly at least 0.05% Cu) solute, such as Cu, Ni or Fe and 0.09 to 0.16% O₂, and consisting of the balance Zr via slight spacings 23 therebetween are arranged in cylindrical channels 11 within the fuel assembly 10 of the boiling water type nuclear reactor, etc. In such a case, the inner sides of the tubes 21 made of the Zr alloy are provided with thin barrier layers 22 consisting of sponge Zr. By which, the corroding and boring of the nuclear fuel rod coating pipes 21 made of the Zr alloy by the cooling water for heat transfer and steam ascending from being in the channels 11 and the occurrence of the accident that the radioactive fission products leak into the cooling water are drastically lessened.

COPYRIGHT: (C)1991,JPO

Description

BACKGROUND OF THE INVENTION (1) Field of the Invention The present invention relates to zirconium-based alloys suitable for use in nuclear reactor applications and, more particularly, to zirconium-based alloys suitable for use in nuclear reactor applications, and more particularly to zirconium-based alloys suitable for use in nuclear reactor applications. Zirconium-based alloys suitable for use as zirconium-based alloys.

(2) Description of Related Art Coatings for fuel elements serve several purposes, but their two primary functions are: The first is to prevent contact and chemical reactions between the nuclear fuel and the coolant or moderator (if a moderator is present) or both (if a coolant and moderator coexist). The second is to prevent radioactive fission products, including gaseous ones, from being released from the nuclear fuel into the coolant or moderator. If the cladding is breached, or the seal is lost, the coolant or moderator and associated systems may become contaminated with long-lived radioactive fission products, thereby interfering with plant operation.

Zirconium-based alloys have long been used as cladding materials for fuel elements for nuclear reactors. Zirconium has a desirable combination of properties in that it has a small thermal neutron cross section and, at the same time, its corrosion resistance in a boiling water reactor environment is at an almost unsatisfactory level. Approximately 12-1.7% tin, 0.07-02% iron, 0.05-0.15% chromium, 0.03-008% nickel, and 0.0515%
Zircaloy-2, an oxygen-containing zirconium alloy, has been widely used in nuclear reactor applications and continues to be used today. While this alloy performs well for nuclear reactor applications, it also has some drawbacks. These shortcomings have stimulated research to find materials that provide improved performance. For example, when a fuel element with a Zircaloy-2 cladding is used in a nuclear reactor, it absorbs hydrogen during reactor operation. When the reactor is shut down and the cladding is cooled, Zircaloy-2 becomes embrittled due to absorbed hydrogen. Zircaloy-4 is a type of alloy developed as a result of research aimed at improving Zircaloy-2. Zirconium-4 is similar to Zirconium-2, but has a lower nickel content [up to 0.007
(weight)%] and slightly higher iron content. Note that Zircaloy-4, which is an improved alloy for Zircaloy-2, was developed for the purpose of reducing hydrogen absorption in Zircaloy-2. Zircaloy-2 and Zircaloy-4 are referred to herein as Zircaloy alloys or Zircaloys.

Zircaloy alloys can be used at nuclear reactor operating temperatures (usually around 290°C) in water where there is no radiation derived from nuclear fission reactions.
) is the best corrosion resistant material when tested under 2
The corrosion rate in water at 90° C. is extremely low, and the corrosion product is a tightly adhered and uniform black ZrO2 layer. However, in actual use, Zircaloy alloy
The instruments are not only irradiated but also exposed to radiolysis products present in the reactor water. Under such conditions, the corrosion resistance of the Zircaloy alloy decreases and its corrosion rate increases.

Research efforts directed toward improving the corrosion resistance of zirconium-based alloys have yielded several results.In some cases, the alloys have been subjected to closely controlled heat treatments either before or after manufacturing the material. However, additional heat treatment cycles generally increase the cost of obtaining the finished product and, if welding is required during installation, There is also the possibility that the area affected by the heat of the welding operation has a different corrosion resistance than the rest of the product.

Furthermore, in an effort to overcome the reduced corrosion resistance of these alloys when subjected to irradiation, it has also been proposed to vary the types of alloying elements and the proportions of alloying elements.

The reduction in corrosion resistance of Zircaloy alloys under actual nuclear reactor conditions is not at all manifested as a -like increase in corrosion rate. Specifically, in addition to the formation of a black ZrO2 layer, localized or nodular corrosion phenomena are sometimes observed on Zircaloy alloy tubes in boiling water nuclear reactors. Such a nodular corrosion reaction is highly undesirable in that it not only increases the corrosion rate but also produces a white ZrO2 bloom that has poorer adhesion and less density than the black Zr02 layer.

The increased corrosion rate resulting from nodular corrosion reactions tends to shorten the service life of the cladding. Additionally, such nodular corrosion reactions have a detrimental effect on the efficient operation of a nuclear reactor. White ZrO2, which has poor adhesion, easily peels off from the tube and mixes into the reactor water. On the other hand, even if the nodular corrosion products do not flake off, if the nodular corrosion products multiply and cover all or most of the pipe with white ZrO2 of low density, heat will be transferred through the pipe. The efficiency with which it is transferred into the water is reduced.

Since it is not possible to use radiation sources to simulate the irradiation that occurs within a nuclear reactor, it is not easy to reproduce actual reactor conditions for routine experimental studies. Moreover, obtaining data from actual use within a nuclear reactor is an extremely time consuming task. For this reason] 0, no conclusive evidence explaining the exact corrosion mechanism leading to nodular corrosion has so far been obtained. As a result, the only way to determine whether other types of alloys were susceptible to nodular corrosion was to actually place test specimens made from the alloys in a nuclear reactor. be.

(Excluding the presence of radiation) When laboratory tests are carried out in water under conditions commonly found in nuclear reactors, i.e. approximately 300°C and 1100 Qpsi, it has been found that Nodular corrosion products such as those seen do not form on Zircaloy alloys.However, at elevated temperatures above 500°C and
Exposure to steam under pressures elevated to 500 psig can produce nodular corrosion products on Zircaloy alloys in laboratory tests, such as those seen on Zircaloy alloys used in nuclear reactors. . In particular, Zircaloy specimens annealed at 750° C. for 48 hours are susceptible to nodular corrosion under such test conditions. That is, a Zircaloy alloy specimen subjected to the above-mentioned annealing was heated for a relatively short period of time (i.e., 24
When tested over an extended period of time, the test results in nodular corrosion comparable to that experienced by Zircaloy cladding used in actual nuclear reactors. These high temperatures and pressures can simulate the environment inside a nuclear reactor, allowing researchers to determine the susceptibility of new alloys to nodular corrosion. Using this test method, specimens of the new alloy and specimens of the Zircaloy alloy can be tested and compared under the same conditions.

New alloys that are considered useful as replacements for Zircaloy alloys must not only have lower susceptibility to nodular corrosion than Zircaloy alloys, but must also have a lower susceptibility than Zircaloy alloys to ensure adequate service life. An equivalent and satisfactory uniform corrosion rate must be maintained. Although Zircaloy alloys have been widely used as fuel rod cladding materials and are known to have a number of desirable properties, there is a need for alternative alloys to also have these properties. Specifically, Zircaloy alloys have a small neutron absorption cross section, are tough, ductile, and extremely stable at temperatures below 750°F, and as mentioned above, have excellent uniformity in water at reactor operating temperatures. It exhibits corrosion resistance.

On the other hand, when examining the performance of fuel elements, it was discovered that brittle failure of the cladding occurs due to the complex interactions between the nuclear fuel, the cladding, and the fission products (produced by the fission reaction). .

Moreover, it has been found that such undesirable performance is caused by localized mechanical stresses on the cladding due to differences in thermal expansion and friction between the nuclear fuel and the cladding. That is, during operation of a nuclear reactor, fission products are generated in the nuclear fuel due to a nuclear fission chain reaction, and these fission products are released from the nuclear fuel and exist on the coating surface. Certain fission products such as iodine and cadmium
The application of local stress or strain in the presence of objects can cause coating failure through a phenomenon known as stress corrosion cracking or liquid metal embrittlement.

SUMMARY OF THE INVENTION The present invention relates to a corrosion resistant fuel element that includes a corrosion resistant zirconium alloy and a cladding vessel constructed from a tube made of such corrosion resistant zirconium alloy. According to one embodiment, about 0
．． 5-2.0% (by weight) tin, about 0.24-0.40
(by weight)% solute and the balance zirconium, the solute consisting of copper, nickel and iron, and the copper content being at least 0.05% (by weight).
Corrosion resistant alloys are provided.

According to another embodiment, about 0.5-2.0% (by weight)
of tin, a solute composed of copper, iron and nickel, each present in a content of 0.05-0.20% (by weight);
and a balance zirconium.

According to yet another embodiment, the composition comprises about 0.5-2.0% (by weight) tin, about 0.25-0.35% (by weight) solute, and the balance zirconium; A third corrosion-resistant alloy is provided, wherein the solute consists of copper and nickel, and the copper content is at least 0.05% (by weight).

These alloys maintain satisfactory uniform corrosion rates when tested with water and steam, and exhibit improved nodular corrosion resistance when exposed to steam at elevated pressures and temperatures.

A corrosion-resistant fuel element is obtained by fabricating an elongated cladding vessel using the first, second, or third zirconium alloy as described in paragraph 1 above.

Improved corrosion resistant fuel elements can also be produced using composite coated vessels, such as those made by metallurgically bonding a surface layer to the outside of a Zircaloy alloy tube.

Such surface layer is comprised of the zirconium alloy No. 1, second or third as described above, and has a thickness equal to about 5 to 20% of the thickness of the zircaloy alloy tube. Such surface layer does not provide a protective shield of sufficient thickness to prevent nodular corrosion effects on the Zircaloy alloy tube.

Additionally, fuel elements may be manufactured using such composite coated vessels that are resistant to nodular corrosion, stress corrosion cracking, and liquid metal embrittlement.

Such a composite coated container is formed by metallurgically bonding a corrosion-resistant surface layer to the outside of a Zircaloy alloy tube and metallurgically bonding a zirconium barrier layer to the inside of the Zircaloy alloy tube. Such internal barrier layer is comprised of medium purity zirconium, such as spongy zirconium, and
It has a thickness equal to about 1-30% of the thickness of the Zircaloy alloy tube.

The outer surface layer, on the other hand, is comprised of the first, second or third zirconium alloy as described above and has a thickness equal to about 5-20% of the thickness of the zircaloy alloy tube.

To produce a coated container made of a zirconium alloy, an extruded billet made of a first, second or third zirconium alloy as described above is heated to about 590-650°C and then formed into a tubular shape by extrusion. . The tube thus obtained is subjected to standard diameter reduction operations and approximately 570-59
Heat treatment at 0° C. provides the desired tube dimensions and mechanical properties. The standard diameter reduction operation for zirconium alloy tubes used in fuel elements is Pilker rolling. Pilger rolling is a diameter reduction operation in which the tube is forged against a fixed mandrel placed inside the tube by using a rotating die that runs over the outside surface of the tube.

To produce a composite coated container, a tube stock of Zircaloy alloy is prepared and an outer tube is placed on the outside thereof. The outer tube is made of the first, second or third zirconium alloy as described above. Such a mass is heated to a temperature within the range of 590-650°C and extruded. In the process, a metallurgical bond is created between the two zirconium alloys. Then a diameter reduction operation and 5
The desired tube dimensions and mechanical properties are obtained by heat treatment at 70-590°C. The above outer tube is
At least, the thickness is such as to provide a surface layer corresponding to about 5 to 20% of the thickness of the Zircaloy alloy tube after diameter reduction.

To produce another type of composite coated container, a tube stock of Zircaloy alloy is provided and an outer tube is placed on the outside thereof. The outer tube is comprised of a first, second or third zirconium alloy as described above. Further, it is placed inside a hollow hollow tube material made of metal for forming a partition layer. This hollow collar is made of medium purity zirconium, such as spongy zirconium.

By heating and extruding such an assembly to 590-650° C., a metallurgical bond is created between the outer surface layer and the tube stock and between the inner partition layer and the tube stock. The desired tube dimensions and mechanical properties are then obtained by a diameter reduction operation and heat treatment at 570-590°C. The outer tube and hollow collar as described above provide an outer surface layer that is at least about 5-20% of the thickness of the Zircaloy tube after diameter reduction and about 1-30% of the thickness of the Zircaloy tube. % of the internal barrier layer.

A nuclear fuel substance is sealed in the above-mentioned coated container and composite coated container so as to leave a gap with the container. In a composite cladding vessel having a barrier layer, the barrier layer protects the alloy tube from the nuclear fuel material held within the vessel and also serves to protect the alloy tube from fission products and debris. Based on its purity,
The barrier layer remains soft during irradiation to reduce local stresses within the fuel element, thereby helping to prevent stress corrosion cracking or liquid metal desorption of the alloy tube.

The use of the first, second, and third zirconium alloys and barrier layers as described above in the fuel elements of the present invention may result in significant neutron absorption increase problems, heat transfer impairment problems, or material incompatibility problems. It won't wake you up.

DETAILED DESCRIPTION OF THE INVENTION The alloy of the present invention exhibits uniform corrosion resistance that is considered sufficient for nuclear reactor applications, and that resistance is approximately comparable to the excellent uniform corrosion resistance of Zircaloy alloys. be. The alloys of the present invention also exhibit improved nodular corrosion resistance.

The addition of tin to zirconium has been practiced prior to the present invention, as exemplified by Zircaloy and other known zirconium-based alloys. The presence of tin is α
Although it stabilizes the zirconium in the mold and thereby primarily contributes to the strength of the alloy, the -like corrosion resistance is also improved somewhat by tin. It has been found that when the tin content is less than about 0.5% (by weight), the uniform corrosion rate of the resulting alloy in water becomes unacceptably high. Also, when the tin content is greater than about 20% (by weight), the resulting alloy exhibits unacceptably high levels of accelerated corrosion in laboratory tests using steam. Therefore, the alloy of the present invention is
About 0.5% to about 2.0% (by weight), preferably about 10% to about 1%
．． It has a tin content of 5% (by weight), and most preferably about 1.5% (by weight). The alloys of the present invention also contain certain additional alloying elements, collectively referred to as "solutes."

The solutes in the alloys of the present invention are distinct from the additional alloying elements found in Zircaloy alloys and primarily contribute to the relative improvement in nodular corrosion resistance.

It should be noted that normal impurities are also present in the alloy of the present invention.

The alloys of the present invention may also optionally contain about 0.09 to 0.1
Note that it may contain as much as 6% (by weight) oxygen.

Many of the commercially available spongy zirconiums used to make zirconium-based alloys, such as those of the present invention, contain small amounts of oxygen;
It is 1300 ppmN degree. In some cases, it may be desirable to increase the oxygen content in the alloy. Increasing oxygen is one way to improve room temperature yield strength.

Thus, the alloys of the present invention can be manufactured with added oxygen if desired, but the addition of oxygen has little or no effect on the corrosion resistance of the alloy.

A number of parameters need to be considered when selecting alloying elements to be added to a zirconium-based alloy for use as a fuel cladding in a boiling water nuclear reactor.

The thermal neutron cross-section of the alloying elements must be relatively small so that the products of the fission reaction can easily pass through the fuel cladding, so that the operating efficiency of the boiling water reactor is as high as possible. It is also necessary to take into account the price of the materials, which should not be unreasonably high. It is also necessary to consider the degree of difficulty in producing a zirconium-based alloy containing the alloying element. Furthermore,
It is also desired that the alloying elements improve the corrosion resistance of zirconium under actual or simulated boiling water reactor conditions.

If an element has ever been considered for nuclear reactor applications, its thermal neutron cross section is generally a known property. Material prices can be ascertained by considering historical price data and making extrapolations as necessary. The method for producing the alloy of the present invention is similar to that for producing conventional zirconium-based alloys, and therefore the ease of production can be easily predicted. A suitable alloy manufacturing method includes a method of arc melting a zirconium-based alloy 1 in which an appropriate amount of alloying elements are sealed in a hollow portion.

After the molten metal thus obtained is cast as an alloy fillet 121~, a final molded article is obtained by subjecting it to several finishing operations.

-i, the most difficult of these parameters to predict is determining whether the alloying element in question contributes to improved corrosion resistance.

Zirconium alloys according to the invention have been found to perform substantially better than Zircaloy-2 in tests to determine nodular corrosion resistance. These alloys also perform well in tests to determine -like corrosion resistance.

Specifically, the first alloy is 0.5 to 2.0 (by weight)
% of tin, about 0.24-0.40% (by weight) of solute and the balance zirconium, and the solute contains copper, nickel and iron, and the content of copper is at least 0.24-0.40% (by weight).
0.05% (by weight). The second alloy contains about 0.05% to 20% (by weight) tin, respectively 0.05% to 20% (by weight) tin.
The third alloy is such that it is composed of copper, iron and nickel present in a content of 0.20% (by weight) i''LfSLfS solute balance zirconium.
2. o (wt)% tin, approximately 025-035 (wt)%
of solute and the remainder zirconium, the solute being of copper and nickel, and the copper content being at least 0.05% (by weight).

Copper, nickel, and iron as solute elements have various properties that are desirable for zirconium-based alloys. Such properties include low thermal neutron cross sections, low cost, ease of alloying, and providing excellent corrosion resistance.

Various alloys according to the invention were tested for -like corrosion resistance and nodular corrosion resistance. As a result of these tests, in alloys with relatively low sensitivity to heat treatment,
It has been found that a dramatic reduction in susceptibility to nodular corrosion can be achieved while retaining approximately the same uniform corrosion resistance as with Zircaloy-2. That is, 0.24 (weight)
Alloys having solute contents in the range of 0.74% to 0.740% (by weight) were tested and found to exhibit superior nodular corrosion resistance compared to the performance of Zircaloy-2.

In addition, alloys containing copper and nickel as solutes are
A significant improvement in nodular corrosion resistance was shown when subjected to a 48 hour annealing at 750° C. which increases the susceptibility of Zircaloy-2 to nodular corrosion. Zirconium alloy tubes undergo several heat treatments during their manufacture. Therefore, zirconium alloys containing copper and nickel as solutes will exhibit improved nodule resistance when subjected to appropriate heat treatment during tube manufacture.

Turning now to FIG. 1, there is shown a partially cut away cross-sectional view of a fuel assembly 1o containing a corrosion resistant fuel made in accordance with the present invention. The bending fuel assembly 10 includes a cylindrical channel 11 of generally square cross section, provided at its upper end with a hanging handle 12 and provided with a north piece at its lower end. (However, fuel assembly 1o
(not shown as the lower part of the figure has been omitted). The upper end of the channel 11 is open at 13, and the lower end of the nosepiece is provided with an opening for the inflow of coolant. A group of fuel elements (or fuel rods) 14 is arranged within the channel 1, and an upper tie plate 15 and a lower tie plate (not shown due to the omission of the lower part).
Supported by Typically, liquid coolant enters through an opening at the lower end of the northpiece, passes upwardly around the fuel element 14, and exits hot through an outlet 13 at the top. At that point, the coolant is partially vaporized in a boiling water reactor, and is not vaporized in a pressurized water reactor.

Both ends of the fuel element] 4 are sealed by end plugs 18 welded to the jacket cap. The end plug 8 may also be provided with a strut 19 to facilitate installation of the fuel element into the fuel assembly. A cavity (or plenum) 20 is provided at one end of the fuel element 14;
This allows longitudinal expansion of the nuclear fuel material and accumulation of debris released from the nuclear fuel material. Inside the cavity 20, a nuclear fuel material holding means 24 made of a helical member is arranged, which is particularly useful for handling the fuel elements and transporting the nuclear fuel material in the axial direction. Helps prevent this.

Such a fuel element 14 provides excellent heat transfer between the cladding vessel 17 and the central core 16 and minimizes parasitic neutron absorption, which sometimes occurs due to high velocity coolant flow. The bow is designed to avoid vibrations.

In FIG. 1, a fuel element (or fuel m) 1.4 produced according to the invention is shown in partial section. Such a fuel element 14 includes a central core 16 of nuclear fuel material constituted by a number of fuel pellets of fissile material and/or fuel parent material disposed inside a cladding vessel 17. , fuel pellets may have various shapes such as cylindrical or spherical.
Other forms (eg, granular) of nuclear fuel material may also be used. It should be noted that the physical form of the nuclear fuel material is not important to the present invention. A variety of nuclear fuel materials can be used, including uranium compounds, plutonium compounds, thorium compounds and mixtures thereof, but preferred is uranium dioxide or a mixture of uranium dioxide and plutonium dioxide.

Next, looking at FIG. 2, the nuclear fuel material forming the central core 16 of the fuel element 4 is surrounded by the cladding vessel F.

A central core material 16 is enclosed within the covering container 17 so as to leave a gap 23 between the central core material 6 and the covering container 17 when used in a nuclear reactor. The coating container 17 is made of an alloy tube 21 made of a corrosion-resistant zirconium alloy. Note that the alloy tube 21 is manufactured using the above-mentioned zirconium alloy No. 1, second or third.

The first, second or third zirconium alloy as described above may also optionally contain from about 0.09 to 116% (by weight) O.
Note that it may also contain oxygen. Many of the commercially available spongy zirconiums used to make zirconium alloys, such as those of the present invention, contain small amounts of oxygen, typically between 800 and 1300 ppm.
That's about it.

In some cases, it may be desirable to increase the oxygen content in the alloy. Increasing oxygen is one way to improve room temperature yield strength. Thus, the alloys of the present invention can be made with added oxygen if desired, but the addition of oxygen has little or no effect on the corrosion resistance of the alloy.

Turning now to FIG. 3, there is shown a corrosion resistant fuel element based on another implementation of the present invention. That is, the nuclear fuel material forming the central core material 16 of the fuel element 14 is
Surrounded by 7.

The central core material 16 is enclosed within the composite covering container ↑7 so as to leave a gap 23 between the central core material 16 and the composite covering container 17 when used in a nuclear reactor. The composite coated container 17 includes an alloy tube 21 manufactured using the first, second, or third zirconium alloy as described above. Bonded to the inner surface of the alloy tube 21 is a metal barrier layer 22 which provides a shield between the alloy tube 21 and the central core 16 (held within the composite envelope). The metal partition layer 22 is made of a material with low neutron absorption (
That is, it is made of medium-purity zirconium and accounts for about 1 to about 30% of the thickness of the composite coated container J7. An example of medium purity zirconium is sponge-like zirconium. Such metal barrier layer 22 prevents the zirconium alloy tube portion of the composite cladding vessel 17 from contacting and reacting with gases and fission products from the nuclear fuel material;
It also prevents the generation of local stresses and strains.

The medium-purity zirconium content of the metal barrier layer 22 is important and serves to impart special properties to the metal barrier layer 22. - Inside the shell, the material of the metal barrier layer 22 contains at least about 1000 ppm and less than about 5000 ppm, preferably about 4200 ppm by weight.
Impurities below pm are present. Among those impurities,
Oxygen is maintained within the range of about 200 ppm to about 200 ppm. All other impurities may be within the normal content range of commercially available zirconium sponge for nuclear reactors.

Turning now to FIG. 4, a corrosion resistant fuel element according to yet another embodiment of the present invention is shown. That is, the nuclear fuel 90 material forming the central core 16 of the fuel element 14 is surrounded by a composite envelope 17 . The center core material 16 is enclosed within the composite covering container 17 so that a gap 23 is left between the center core material 16 and the composite covering container 17 when used in a nuclear reactor. Such a composite coated container 17 includes an alloy tube 30 made of Zircaloy alloy.
Contains.

A metal surface layer 32 is bonded to the outer surface of the alloy tube 30 and provides a protective shield for the alloy tube 30 to prevent corrosion. The metal surface layer 32 is comprised of the first, second, or third zirconium alloy described above and has a thickness equal to about 5-20% of the thickness of the alloy tube 30. Such a metal surface layer 32 serves to prevent nodular corrosion of the Zircaloy alloy tube portion of the composite coated vessel 17.

Turning now to FIG. 5, there is shown a corrosion resistant fuel element according to yet another embodiment of the present invention. That is, the nuclear fuel material enclosing the central core 16 of the fuel element 14 is surrounded by the composite cladding vessel 17. The center core material 16 is enclosed within the composite covering container 17 so that a gap 23 is left between the center core material 16 and the composite covering container 17 when used in a nuclear reactor. The composite coated container 17 includes an alloy tube 30 made of Zircaloy alloy.

Bonded to the inner surface of the alloy tube 30 is a metal barrier layer 22 that provides a shield between the alloy tube 30 and the central core 16 (held within the composite jacket). Metal barrier layer 22 is comprised of a material with low neutron absorption (ie, medium purity zirconium as described above) and has a thickness equal to about 1 to about 30% of the thickness of alloy tube 30. Such metal barrier layer 22 prevents the Zircaloy alloy tube portion of composite cladding vessel 17 from contacting and reacting with scum and fission products from the nuclear fuel material, and also prevents the generation of local stress and strain. On the other hand, a metal surface layer 32 is bonded to the outer surface of the alloy tube 30. Metal surface layer 3
2 is made of the first, second, or third zirconium alloy as described above, and is about 5 to 20% of the thickness of the alloy tube 30.
has a thickness equal to . Such a metal surface layer 32 serves to prevent nodular corrosion of the Zircaloy alloy tube portion of the composite coated vessel 17.

The spongy zirconium that constitutes the metal barrier layer in the above composite coated container has a high degree of resistance to radiation hardening, so even after long-term irradiation, the metal barrier layer maintains the desired yield strength and hardness. Structural properties can be maintained at significantly lower levels than with conventional zirconium alloys. In fact, such metal barrier layers do not harden as normal zirconium alloys when subjected to irradiation.

As a result, the metal barrier layer, which has an inherently low yield strength, exhibits plastic deformation during transient power changes, thereby relieving pellet-induced stresses in the fuel element. Pellet-induced stresses in the fuel element can be caused, for example, by the expansion of nuclear fuel pellets into contact with the cladding at reactor operating temperatures (300-350<0>C).

Furthermore, the thickness of the spongy zirconium metal barrier layer bonded to the alloy tube is about 5 to 15 times the thickness of the composite jacket.
%, and it has also been found that it is particularly preferred if it is equal to 10% of the thickness of the composite coated container. In such a case, a reduction in stress will be achieved and a barrier effect sufficient to prevent failure of the composite coated container will be achieved.

The corrosion-resistant coated container used in the fuel element of the present invention can be manufactured using a billet made of the above-mentioned zirconium alloy No. 1, second or third. The billet is heated to 590-650'C and extruded. By subjecting the thus extruded tube to a conventional diameter reduction operation, a coated container having the desired dimensions can be obtained.

According to another method, a hollow collar of spongy zirconium selected to form a metal barrier layer is placed within a hollow billet of a first, second or third zirconium alloy as described above. inserted. Such a mass is heated to a temperature within the range of 590-650°C and extruded.

By subjecting the thus extruded tube to a conventional diameter reduction operation, a composite coated container having the desired dimensions is obtained.

3 4 According to yet another method, a tube stock made of Zircaloy alloy is provided and an outer tube is placed outside it. The outer tube is made of a first, second or third zirconium alloy as described above. Such an aggregate is 59
It is heated to a temperature within the range of 0-650'C and extruded. By subjecting the thus extruded tube to a conventional diameter reduction operation, a composite coated container having the desired dimensions is obtained.

According to yet another method, a tube stock made of Zircaloy alloy is provided and an outer tube is placed outside it. The outer tube is made of the first, second or third zirconium alloy described above. A hollow collar of spongy zirconium selected to form a metal barrier layer is also inserted inside the tube stock. The mass is heated to a temperature within the range of 590-650°C and extruded. By subjecting the thus extruded tube to a conventional diameter reduction operation, an fS composite coated container having the desired dimensions is obtained.

An intermediate annealing and a final annealing are performed during the diameter reduction operation as described above. The grilled Nayoshi is 570-5
It is carried out at a temperature within the range of 90°C.

The present invention also provides (1) a zirconium alloy, (2) a zirconium alloy and an internal barrier layer, (3) a zircaloy alloy and an external surface layer, or (4) a zircaloy alloy,
Also included is a method of manufacturing a fuel element using a cladding or a composite cladding comprising either an external surface layer and an internal barrier layer.

A core of nuclear fuel material is filled into a jacketed container open at one end, leaving a gap therebetween and a void at the open end of the container. A nuclear fuel holding means is then inserted into said cavity and a closing means is placed at the open end of the cladding vessel. As a result, said cavity is kept in communication with the nuclear fuel material. Thereafter, the open end of the coated container is joined to the closure means to form a hermetic seal therebetween.

The present invention provides several advantages, such as extending the service life of the fuel element. That is, improved nodular corrosion resistance protects the strength and integrity of the coated vessel. Composite coated containers with bulkhead layers reduce chemical interactions on the container, reduce local stress on the zirconium alloy tubing portion of the container, and reduce stress corrosion and cracking in the zirconium alloy tubing portion of the container. Reducing fillet corrosion reduces the likelihood of rupture in the zirconium alloy tube section. The present invention also suppresses nuclear fuel material from expanding and directly contacting the zirconium alloy tube, resulting in
It also reduces the generation of local stress on the alloy tube, the initiation or acceleration of stress corrosion of the alloy tube, and the adhesion of nuclear fuel material to the alloy tube.

An important characteristic of the composite coated container of the present invention is that the above improvements are achieved without substantial increases in neutron absorption. Furthermore, the composite coated container of the present invention causes extremely few heat transfer problems. This is because there is no insulation layer to impede heat transfer as there would be if a separate foil or liner were inserted within the fuel element. Furthermore, the composite coated containers of the present invention can be tested by conventional non-destructive testing methods at various stages of manufacture and use.

Examples are provided below to illustrate the improved nodular corrosion resistance of the zirconium alloys used in the present invention.

Example 1 In Table 1, the last three alloys are comprised of Zircaloy-2 in three different cold rolled and heat treated conditions:
Various alloys according to the invention are shown. --like corrosion resistance under conditions equivalent to nuclear reactor operating conditions (excluding the radiation source), i.e., a temperature of 288 °C and a temperature of 15 °C.
The above alloys were tested in water containing s ppm oxygen using a pressure of 1,000 psig.

As can be seen from the results shown in Table 1, the specimens made of the alloy of the present invention exhibited excellent uniform corrosion resistance.

That is, despite being tested over a much longer period of time, the corrosion rate of specimens made of the alloy of the present invention was lower than that of Zircaloy.
The corrosion rate was the same as that of the No. 2 test piece.

7 8 In testing under these conditions, either The specimens also showed no evidence of nodular corrosion products.

Example 2 Table 2 shows the results of tests carried out to determine the susceptibility of the alloys of the invention to nodular corrosion. Such testing was conducted by exposure to steam at 510° C. and 1500 psig. These test conditions produced the same nodular corrosion products (750
Conditions such as those produced in the laboratory on Zircaloy alloys (annealed for 48 hours at °C). For comparison, an annealed Zircaloy alloy tested under the same test conditions would have a weight gain of several thousand mg/kg.
It was about dm2.

The tests in Table 2 were also conducted at various cold rolling and heat treating conditions. The results shown in Table 2 show that the corrosion resistance of these alloys is not significantly dependent on the heat treatment conditions of the specimens. However, it is preferable to heat treat alloys containing copper and nickel as solutes. Some alloys were tested both unannealed and annealed using specimens made from cold rolled sheets. Also, regarding the two types of alloys,
Tests were conducted using only cold rolled and annealed specimens. In each case, the tested alloys of the invention were annealed at 750'C for 48 hours, a heat treatment that removed the nodular corrosion resistance from Zircaloy-2 during laboratory tests using steam. It's like leaving.

All weight increases shown in Table 2 are at (750'C
The results are much better than those obtained when testing Zircaloy alloys that have been sensitized (by annealing for 48 hours). Many of the inventive alloys tested were 100 mg/+Im
The remaining alloy also showed a weight increase of less than 107
It showed a weight increase of mg/dm2. In contrast, when a sensitized Zircaloy alloy is tested under the same test conditions, the weight gain obtained is several thousand mg/dm2, as mentioned above.
It's just a matter of degree.

In addition to the reduced weight gain demonstrated in the present alloys, none of these alloys showed evidence of nodular corrosion products. It is thus clear that these alloys exhibit improved nodular corrosion resistance under the test conditions described in section 2.

3 44

[Brief explanation of drawings]

FIG. 1 is a partially cutaway cross-sectional view of a fuel assembly including a fuel element manufactured in accordance with one embodiment of the present invention, and FIG. 2 is an enlarged cross-sectional view of the combustion element taken along line 2--2 in FIG. 3-5 are enlarged cross-sectional views of three fuel elements made in accordance with other embodiments of the present invention. In the figure, ] 0 is the fuel assembly, 11 is the channel, ] 4 is the fuel element, 15 is the upper tie plate, 16 is the center core material, 17
is a covering container or a composite covering container, 18 is a closing means or an end plug, 1 is a column, 20 is a cavity, 2 is a zirconium alloy tube, 22 is a metal partition layer, 23 is a gap, 24 is a nuclear fuel material holding means, 30 represents a Zircaloy alloy tube, and 32 represents a metal surface layer. JP-A-3 64427 (14) FIG.

Claims (1)

[Claims] 1. about 0.5-2.0% (by weight) tin, about 0.24-2.0% tin;
Corrosion resistance, characterized in that it consists of 0.40% (by weight) solute and the balance zirconium, said solute consists of copper, nickel and iron, and the content of copper is at least 0.05% (by weight) alloy. 2. Corrosion-resistant alloy according to claim 1, additionally containing 0.09 to 0.16% (by weight) of oxygen. 3. about 0.5-2.0% (by weight) tin, each 0.
Corrosion-resistant alloy, characterized in that it consists of a solute composed of copper, iron and nickel, present in a content of 0.05 to 0.20% (by weight), and the remainder zirconium. 4. Corrosion-resistant alloy according to claim 3, additionally containing 0.09 to 0.16% (by weight) of oxygen. 5. Approximately 0.5 to 2.0% (by weight) tin, approximately 0.25 to 2.0% (by weight)
0.35% (by weight) of solute and balance zirconium, said solute consisting of copper and nickel;
and a copper content of at least 0.05% (by weight). 6. Corrosion-resistant alloy according to claim 5, additionally containing 0.09 to 0.16% (by weight) of oxygen. 7. (a) 0.5-2.0% (by weight) tin, 0.24
an elongated coated vessel constructed from an alloy tube made of a zirconium alloy consisting of ~0.40% (by weight) solute and the balance zirconium, said solute consisting of copper, nickel and iron, (b ) consisting of nuclear fuel material selected from the group consisting of uranium compounds, plutonium compounds, thorium compounds and mixtures thereof;
(c) a central core material disposed within the container to partially fill the container, leaving a gap between the container and a void at one end of the container; A corrosion-resistant fuel element, characterized in that it comprises closure means integrally and sealingly secured thereto at each end, and (d) elements of nuclear fuel material retention means disposed within said cavity. 8. The zirconium alloy is 0.5 to 2.0% (by weight)
of tin, a solute composed of copper, iron and nickel, each present in a content of 0.05-0.20% (by weight);
8. The corrosion resistant fuel element of claim 7, further comprising zirconium and the balance zirconium. 9. The zirconium alloy is 0.5 to 2.0% (by weight)
of tin, 0.25-0.35% (by weight) of solute and the balance zirconium, said solute consisting of copper and nickel, and the copper content is at least 0.05%.
% (by weight). 10. A composite coated vessel in which the coated vessel is formed by metallurgically bonding a partition layer of sponge-like zirconium to the inner surface of the alloy tube, wherein the thickness of the partition layer is approximately equal to the thickness of the alloy tube. Corrosion resistant fuel element according to claim 7, wherein the corrosion resistant fuel element is equal to 1 to 30%. 11. The corrosion-resistant fuel element of claim 7, wherein said zirconium alloy additionally contains 0.9-0.16% (by weight) oxygen. 12. (a) An alloy tube made of Zircaloy alloy, and a tube connected to the outside of the alloy tube and having a thickness of about 5 to 20% of the alloy tube.
a surface layer of a zirconium alloy having a thickness equal to 0.5 to 2.
0% (by weight) tin, 0.24-0.40% (by weight) solute and the balance zirconium, said solute consisting of copper, nickel and iron, and the copper content being at least 0.05% (by weight). (b) comprising a nuclear fuel material selected from the group consisting of uranium compounds, plutonium compounds, thorium compounds and mixtures thereof, leaving a gap between said container and said container; (c) a central core disposed within the container, partially filling the container, leaving a void at one end of the container; (c) a seal integrally attached to each end of the container; Corrosion-resistant fuel element, characterized in that it consists of closure means fixed in a stationary state, and (d) elements of nuclear fuel material retention means arranged in said cavity. 13. The zirconium alloy is 0.5 to 2.0 (weight)
13. Corrosion-resistant fuel element according to claim 12, consisting of a solute composed of copper, iron and nickel present in a content of 0.05% to 0.20% (by weight) each, and the balance zirconium. 14. The zirconium alloy is 0.5 to 2.0 (weight)
% tin, 0.25-0.35% (by weight) solute and the balance zirconium, said solute consisting of copper and nickel, and the copper content being at least 0.0%.
13. The corrosion resistant fuel element of claim 12, wherein the corrosion resistant fuel element is 5% (by weight). 15. The corrosion resistant fuel element of claim 12, wherein said zirconium alloy additionally contains 0.9-0.16% (by weight) oxygen. 16. The composite coated container has a spongy zirconium barrier layer metallurgically bonded to the inner surface of the alloy tube, and the barrier layer has a thickness of about 1 the thickness of the alloy tube.
13. The corrosion resistant fuel element of claim 12, wherein the corrosion resistant fuel element is equal to -30%.